psr-0315 |
AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |
ccc-0082 |
ANISN-E, 1-D Transport Program ANISN with Exponential Model |
nea-0363 |
ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration |
ccc-0254 |
ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering |
ccc-0255 |
ANISN-W, 1-D Transport Calculation for Deep Penetration Problems |
ccc-0514 |
ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering |
ccc-0459 |
BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup |
nea-1678 |
BOT3P5.4, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results |
nesc0387 |
CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search |
ccc-0643 |
CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC |
ccc-0726 |
CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System |
ccc-0829 |
COG11.1, Multiparticle Monte Carlo Code System for Shielding and Criticality Use |
iaea1226 |
CORD-2, PWR Core Design and Fuel Management |
nea-1903 |
CRISTAL V2.0.3, Criticality calculation package |
ccc-0547 |
DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport |
ccc-0649 |
DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method |
ccc-0784 |
DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems |
nea-0391 |
DLS, 2-D Diffusion with Line-of-Sight Method for Cavities |
ccc-0650 |
DOORS3.2A, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport |
ccc-0276 |
DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling |
ccc-0320 |
DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature |
uscd1234 |
DRAGON 3.05D, Reactor Cell Calculation System with Burnup |
nea-0322 |
DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method |
nea-1683 |
ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses |
nesc0156 |
EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
nea-0443 |
FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry |
nea-0545 |
FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method |
nea-0896 |
FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method |
nesc0380 |
GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR |
iaea1271 |
GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion |
nesc0277 |
HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation |
nea-0624 |
JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |
ccc-0510 |
KENO-IV(RG), KENO-IV with Random Geometry |
ccc-0548 |
KENO5A-PC, Monte-Carlo Criticality with Supergrouping |
uscd1241 |
MCART, solve the time dependent neutron transport equation |
nea-1643 |
MCB1C, Monte-Carlo Continuous Energy Burnup Code |
nea-1733 |
MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials |
iaea0889 |
MCRAC/RBI, In Core Fuel Management, Program of PFMP System |
ccc-0841 |
MMS3D, Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness |
nea-0527 |
MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method |
nea-1905 |
MORET 5.D.1, Monte Carlo simulation tool to solve transport equation for neutrons |
nea-1633 |
MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers |
nea-1896 |
MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA |
nea-1673 |
MVP/GMVP V.3, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods |
ccc-0641 |
NESTLE 5.2.1, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM) |
nea-1591 |
OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo |
nea-1324 |
OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |
ccc-0760 |
PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code |
ccc-0842 |
PARTISN 8.29, Time-Dependent, Parallel Neutral Particle Transport Code System |
ccc-0708 |
REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |
nea-1840 |
SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |
nea-1923 |
SERPENT V2.2.X, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |
iaea1437 |
SUPERMC 3.3.0, Super Monte Carlo simulation program for nuclear and radiation process |
ccc-0204 |
SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |
ccc-0638 |
TART2022, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code |
nesc0558 |
TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron |
ccc-0759 |
TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System |
ccc-0543 |
TORT-DORT, 1-D 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
nea-1716 |
TRIPOLI-4 VERS. 8.1, 3D general purpose continuous energy Monte Carlo Transport code |
nea-1878 |
TRIPOLI-4 version 9S, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation |
nea-1086 |
TRISTAN, 3-D fixed source radiation transport |
nea-0415 |
TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search |
uscd1239 |
VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling |
ccc-0654 |
VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup |
nea-1856 |
VESTA 2.1&AURORA1.0, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool |
ccc-0754 |
VIM 5.1, Steady-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections |
iaea0871 |
VPI-NECM, Nuclear Engineering Program Collection for College Training |
nea-0655 |
VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
iaea1440 |
VSOP99-11, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
nea-1882 |
XSUN-2023, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |
nea-1206 |
ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure |
nea-1264 |
ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis |